• 한국기계가공학회지
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• 제14권6호
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• pp.21-26
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• 2015

This study presents a development of the cleaning system in a nuclear power plant condenser. The tube cleaning system is very important equipment in a power plant condenser. Specially, removal of the fouling is a key process in the condenser tube. The objective of this study is development of a ball collector system for cleaning a condenser by using a sponge ball. This study uses CFD in order to optimize design of the ball strainer screen. Through the CFD, the implication of the ball strainer screen for static pressure distribution is examined. Results of research, this study have developed a 1/5 scale model for application to the power plant and developed a performance test equipment.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.1053-1054
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• 2005

• 대한전기학회:학술대회논문집
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• 대한전기학회 1993년도 하계학술대회 논문집 A
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• pp.525-528
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• 1993

KEPCO Research Center is carrying out a simulator(full scope replica type) development project for two nuclear power plants(Kori-2, Younggwang-3,4) and one fossil power plant(Poryong-3,4). In this project, we aim not only the installation of high performance simulators at the power plant sites but also the realization of self reliance of power plant simulation technology in Korea. In the course of preparing procurement specification for the 3 simulators, the present status of power plant simulation technology has been surveyed and is presented in this paper. The fidelity of simulation and the automation of simulation model production has been greatly improved due to the ever increasing computing power of today's workstations. The need and importance of the application of high fidelity simulators to the operator training is refocused since the accident at TMI Nuclear Power Plant, U.S.A.

• Nuclear Engineering and Technology
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• 제28권3호
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• pp.265-273
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• 1996

Among those tests performed during the Yonggwang Nuclear Power Plant Units 3 and 4 (YGN 3&4) Power Ascension Test period, the Loss of a Main Feedwater Pump test at l00% power is one of the major test which characterize the capability of YGN 3&4. In this event, one of the two normally operating main feedwater pumps is tripped resulting in a 50% reduction in the feedwater flow. Unless the NSSS and Turbine/Generator control systems actuate properly, the reactor will be tripped on low SG water level or high pressurizer pressure. The test performed at Unit 3 was successful by meeting all acceptance criteria, and the plant was stabilized at a reduced power level without reactor trip. The measured test data for the major plant parameters are compared with the predictions made by the KISPAC computer code, an updated best-estimate plant performance analysis code, to verify and validate its applicability. The comparison results showed good agreement in the magnitude as well as the trends of the major plant parameters. Therefore, the KISPAC code can be utilized for the best-estimate nuclear power plant design and simulation tool after a further verification using other plant test data.

• 산업경영시스템학회지
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• 제27권1호
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• pp.109-120
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• 2004

In large scale complex system such as a nuclear power plant, it is important to select guidelines and/or checklist to evaluate the system performance, especially human performance for visual information while the number of evaluation items of the guidelines and checklist is voluminous. This paper presents the methodology and experiment for the relative weights or priority selection of evaluation items on the advanced information display of main control room in a nuclear power plant. To summarize this, 1) many human factors guidelines of Visual Display Terminal(VDT) displays are collected, 2) the collected guidelines are integrated and unified based on some rules in a way to avoid confusion or errors about work performances of operator groups, 3) using the unified guidelines, the more important items are defined when the advanced information indexes are applied by using the Analytic Hierarchy Process(AHP). For employing the AHP, the decisions and response of many human factors evaluation specialists in this field are collected to get the priority order of the evaluation items of VDT. The result of this paper will be applied for the evaluation of the usability of next generation of nuclear power plant which is focused on the visual information display on VDT.

• 한국안전학회지
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• 제36권5호
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• pp.79-85
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• 2021

A standard engineering model that reflects the current organization system and engineering operation process of domestic nuclear power plants was developed based on the Standard Nuclear Performance Model developed by the American Nuclear Energy Association. The level 0 screen, which is the main screen of the engineering model computer system, consisted of an object tree structure, which provided information that is phased down from a higher structure level to a lower structure level (i.e., level 3). The level 1 screen provided information related to the sub-process of the engineering operation, whereas the Level 2 screen provided information related to each engineering operation activity. In addition, the Level 2 screen provided additional functions, such as linking electronic procedures/guidelines, providing electronic performance forms, and connecting legacy computer systems (such as total equipment reliability monitoring system, configuration management systems, technical information systems, risk monitoring systems, regulatory information, and electronic drawing system). This screen level increased the convenience of user's engineering tasks by implementing them. The computerization of an engineering model that connects the entire engineering tasks of an establishment enables the easy understanding of information related to the engineering process before and after the operation, and builds a foundation for the enhancement of the work efficiency and employee capacity. In addition, KHNP developed an online training module, which operates as an e-learning process, on the overview and utilization of a standard engineering model to expand the understanding of standard engineering models by plant employees and to secure competitiveness.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2928-2938
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• 2020

A light water nuclear Reactor has been exergy analyzed, and the rate of irreversible exergy loss and exergy destruction is calculated for each of its components. The ratio of these losses compared to the total input exergy loss is calculated, which shows that most irreversible losses occur in the reactors, turbines, steam generators, respectively, as well as in the downstream operations. The main aim of this paper is to optimize the power plant using an innovative firefly algorithm and then to propose a novel strategy to improve the overall performance of the plant. As shown in the results, the exergy destruction rate of the plant decreased by 1.18% using the firefly method, and the exergy efficiency of the plant reached 29.3% comparing to the operational amount of 28.99%. Also, the results of the firefly optimization process compared to the Genetic algorithm and gravitational search algorithm to study the accuracy of the model for exergy analysis fitness problems in the power plants and the results of this comparison has shown that the results are nearly similar in the mentioned methods. However, the firefly is faster and more accurate in limited iterations.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2652-2659
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• 2020

The methyl iodide (CH₃I) removal performance of tri-ethylene-di-amine impregnated activated carbon (TEDA-AC) used in the air cleaning unit of nuclear power plants (NPPs) should be maintained at least 99% between 24 month-performance test period. In order for evaluating the effectiveness of TEDA-AC on the removal performance of CH₃I in nuclear power plant during the operation of NPPs, the long-term test for up to 15 months was carried out under the simulated operating conditions (e.g., 25 ℃, RH 50%, ppb level poisoning gases injection) at nuclear power plants (NPPs). The TEDA-AC samples were analyzed with the Brunauer-Emmett-Teller (BET) specific surface area and TEDA content as well as CH₃I penetration test. It is clearly evident that more than 99% of CH₃I removal performance of TEDA-AC was observed in the TEDA-AC samples during 15 months of long-term operation under the simulated NPP operating conditions including the ppb level of organic and oxide form of poisoning gases. BET specific surface area and TEDA content that can affect the CH₃I removal performance of TEDA-AC were also maintained as those in new TEDA-AC during 15 months of long-term operation.

• 시스템엔지니어링학술지
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• 제6권1호
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• pp.47-52
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• 2010

Human Factors Engineering (HFE) for Main Control Room (MCR) of Nuclear Power Plant (NPP) has been applied to optimize the design and operation of Man-Machine Interface (MMI) between operators and their equipment in consideration of physical, psychological and cognitive aspects. However, it has been observed that operators complain about environmental discomfort in the MCR since the operators in the MCR experience excessive stress due to the environmental factors such as inappropriate interior and lighting system. Since the HFE is an essential factor for the high fidelity performance of operators in the MCR, the adequate MCR environment design with HFE rules and guidelines is as much important to enhance the operability and reliability of the MCRs. Therefore, there has been a strong need to design a pleasant environment for the MCR to improve human performance of the operators.

• 한국안전학회지
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• 제25권2호
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• pp.24-28
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• 2010

The nuclear power Plant onsite AC electrical power sources are required to supply power to the engineering safety facility buses if the offsite power source is lost. Typically, Diesel Generators are used as the onsite power source. The 125 VAC buses are part of the onsite Class 1E AC and DC electrical power distribution system. The DC power distribution system ensure the availability of DC electrical power for system required to shutdown the reactor and maintain it in a safety condition after an anticipated operational occurrence or a postulated Design Base Accident. Recently, onsite DC power supply system trip occurs the loss of system function. To obtain the performance such as reliability and availability, we analyzed the cause of battery charger trip and described the improvement of DC power supply system reliability. Finally, we provide reliability performance criteria of charger in order to ensure the probabilistic goals for the safety of the nuclear power plants.